ML110270190
| ML110270190 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 01/27/2011 |
| From: | Rich D NRC/RGN-II/DRP/RPB3 |
| To: | Franke J Progress Energy Florida |
| References | |
| IR-10-005 | |
| Download: ML110270190 (38) | |
See also: IR 05000302/2010005
Text
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION II
245 PEACHTREE CENTER AVENUE NE, SUITE 1200
ATLANTA, GEORGIA 30303-1257
January 27, 2011
Mr. Jon A. Franke
Vice President, Crystal River Nuclear Plant
Crystal River Nuclear Plant (NA2C)
15760 W. Power Line Street
Crystal River, FL 34428-6708
SUBJECT: CRYSTAL RIVER UNIT 3 - NRC INTEGRATED INSPECTION REPORT
Dear Mr. Franke:
On December 31, 2010, the US Nuclear Regulatory Commission (NRC) completed an
inspection at your Crystal River Unit 3. The enclosed inspection report documents the
inspection findings, which were discussed on January 10, 2011, with you and other members of
your staff.
The inspection examined activities conducted under your license as they related to safety and
compliance with the Commissions rules and regulations and with the conditions of your license.
The inspectors reviewed selected procedures and records, observed activities, and interviewed
personnel.
Based on the results of this inspection one licensee identified violation, which was of very low
safety significance, is listed in Section 4OA7 of the report. If you contest the non-cited violation,
you should provide a response within 30 days of the date of this inspection report, with the basis
for your denial, to the Nuclear Regulatory Commission, ATTN.: Document Control Desk,
Washington DC 20555-001; with copies to the Regional Administrator Region II; the Director,
Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-
0001; and the NRC Resident Inspector at the Crystal River Unit 3 site.
FPL 2
In accordance with 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter and its
enclosure will be available electronically for public inspection in the NRC Public Document
Room or from the Publicly Available Records (PARS) component of the NRCs document
system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-
rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
Daniel W. Rich, Chief
Reactor Projects Branch 3
Division of Reactor Projects
Docket No. 50-302
License No. DPR-72
Enclosure: Inspection Report 05000302/2010005
w/Attachment: Supplemental Information
cc w/encl: (See page 3)
___ML110270190______________ G SUNSI REVIEW COMPLETE
OFFICE RII:DRP RII:DRP RII:DRS RII:DRS RII:DRS RII:DRS RII:DRP
SIGNATURE TXM1 via email RJR1via email DRich for RKH1 via email WTL via email DRich for SRS5 via email
NAME TMorrissey RReyes RChou RHamilton WLoo CDykes SSandal
DATE 01/21/2011 01/21/2011 01/27/2011 01/21/2011 01/25/2011 01/27/2011 01/21/2011
E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO YES NO YES NO
OFFICE RII:DRS RII:DRP RII:DRS RII:DRS RII:DRS
SIGNATURE DRich for WXD1via email RSB2 via email MKM3 via email LFL via email
NAME PHiggins WDeschaine RBaldwin MMeeks LLake
DATE 01/27/2011 01/21/2011 01/27/2011 01/21/2011 01/24/2011
E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO YES NO YES NO
OFFICE RII:DRS RII:DRP RII:DRP RII:DRS
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NAME RWilliams SNinh DRich RAiello
DATE 01/20/2011 01/21/2011 01/27/2011 01/26/2011
E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO YES NO YES NO
FPC 3
cc w/encl: Joseph W. Donahue
R. J. Duncan, II Vice President
Vice President Nuclear Oversight
Nuclear Operations Carolina Power and Light Company
Carolina Power & Light Company Electronic Mail Distribution
Electronic Mail Distribution
Jack E. Huegel
Brian C. McCabe Manager, Nuclear Oversight
Manager, Nuclear Regulatory Affairs Crystal River Nuclear Plant
Progress Energy Carolinas, Inc. Electronic Mail Distribution
Electronic Mail Distribution
David T. Conley
James W. Holt Associate General Counsel
Plant General Manager Legal Dept.
Crystal River Nuclear Plant (NA2C) Progress Energy Service Company, LLC
Electronic Mail Distribution Electronic Mail Distribution
Stephen J. Cahill Mark Rigsby
Engineering Manager Manager, Support Services - Nuclear
Crystal River Nuclear Plant (NA2C) Crystal River Nuclear Plant (NA2C)
Electronic Mail Distribution Electronic Mail Distribution
R. Alexander Glenn Attorney General
Associate General Counsel Department of Legal Affairs
(MAC - BT15A) The Capitol PL-01
Florida Power Corporation Tallahassee, FL 32399-1050
Electronic Mail Distribution
Ruben D. Almaguer
Christos Kamilaris Director
Director Division of Emergency Preparedness
Fleet Support Services Department of Community Affairs
Carolina Power & Light Company Electronic Mail Distribution
Electronic Mail Distribution
Chairman
William A. Passetti Board of County Commissioners
Chief Citrus County
Florida Bureau of Radiation Control 110 N. Apopka Avenue
Department of Health Inverness, FL 36250
Electronic Mail Distribution
Daniel R. Westcott
Supervisor
Licensing & Regulatory Programs
Crystal River Nuclear Plant (NA1B)
Electronic Mail Distribution
FPC 4
Letter to Jon Franke from Daniel Rich dated January 27, 2011
SUBJECT: CRYSTAL RIVER UNIT 3 - NRC INTEGRATED INSPECTION REPORT
Distribution w/encl:
C. Evans, RII
L. Douglas, RII
OE Mail
RIDSNRRDIRS
PUBLIC
RidsNrrPM Crystal River Resource
U.S. NUCLEAR REGULATORY COMMISSION
REGION II
Docket No.: 50-302
License No.: DPR-72
Report No.: 05000302/2010005
Licensee: Progress Energy (Florida Power Corporation)
Facility: Crystal River Unit 3
Location: Crystal River, FL
Dates: October 1, 2010 - December 31, 2010
Inspectors: T. Morrissey, Senior Resident Inspector
R. Reyes, Resident Inspector
R. Chou, Reactor Inspector (Section 4OA5.2)
R. Hamilton, Senior Health Physicist (Sections 2RS6, 40A1.2)
W. Loo, Senior Health Physicist (Sections 2RS1, 40A1.2, 40A5.4)
C. Dykes, Health Physicist (Sections 2RS8)
S. Sandal, Senior Reactor Inspector (Section 4OA5.2 and 4OA5.3)
P. Higgins, Senior Reactor Inspector (Section 4OA5.3)
W. Deschaine, Reactor Inspector (Section 4OA5.3)
R. Baldwin, Senior Operations Engineer (Section 4OA5.5)
M. Meeks, Operations Engineer (Section 4OA5.5)
L. Lake, Senior Reactor Inspector (Section 4OA5.2)
G. Thomas, Structural Engineer (Section 4OA5.2)
F. Farhad, Senior Structural Engineer (Section 4OA5.2)
R. Williams, Reactor Inspector (Section 4OA5.6)
R. Aiello, Senior Operations Engineer (Section 1R11)
Approved by: D. Rich, Chief,
Reactor Projects Branch 3
Division of Reactor Projects
Enclosure
SUMMARY OF FINDINGS
IR 05000302/2010005; 10/01/2010 -12/31/2010; Crystal River Unit 3; Routine Integrated
Report.
The report covered a three month period of inspection by resident inspectors, regional
operations engineers, regional reactor inspectors, headquarters inspectors, and regional health
physicists. The NRCs program for overseeing the safe operation of commercial nuclear power
reactors is described in NUREG-1649, Reactor Oversight Process, Revision 4, dated
December 2006.
A. NRC Identified & Self-Revealing Findings
No findings were identified
B. Licensee Identified Violations
One violation of very low safety significance, which was identified by the licensee, has
been reviewed by the inspectors. Corrective actions taken or planned by the licensee
have been entered into the licensees corrective action program. The violation and
corrective action tracking number is listed in Section 4OA7 of this report.
Enclosure
REPORT DETAILS
Summary of Plant Status:
Crystal River 3 began the inspection period with the full core off-loaded to the spent fuel pool.
On November 19, 2010, the licensee commenced transfer of reactor fuel assemblies to the
reactor vessel. On November 27, 2010, after fuel reload was complete and the reactor head
fully tensioned, the unit entered Mode 5. Unit 3 remained in Mode 5 for the remainder of the
inspection period.
1. REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity
1R01 Adverse Weather Protection
.1 Seasonal Susceptibility: Cold Weather Preparation
a. Inspection Scope
The inspectors evaluated the licensees readiness for mitigating cold weather to assure
that vital systems and components were protected from freezing in accordance with the
licensees administrative instruction AI-513, Seasonal Weather Preparations, Section
4.1, Cold Weather Preparations. The inspectors walked down portions of the
systems/areas listed below to check for any unidentified susceptibilities. Operability of
heat trace circuits and set points of temperature controls was verified. Nuclear condition
reports (NCRs) were reviewed to verify that the licensee was identifying and correcting
cold weather protection issues.
- Alternate AC emergency diesel generator EGDG-1C building
- EGDG-1A and 1B rooms
- Emergency feed water pump EFP-3 building including EFP-3 heat tracing
- Intermediate building 95 elevation EFP-1 and EFP-2 areas
b. Findings
No findings were identified.
.2 Site Specific Weather Condition
a. Inspection Scope
During the period listed below, the inspectors verified that the licensee implemented
Administrative Instruction AI-513, Seasonal Weather Preparations, Sections 4.2
(Freezing Weather) and/or 4.3 (Freezing Weather Monitoring). The inspectors walked
down portions of the A and B emergency diesel generator (EGDG) systems; the
alternate AC diesel generator system; and the EFP-3 building to check for any
unidentified susceptibilities to cold weather. Nuclear condition reports were reviewed to
Enclosure
4
verify that the licensee was identifying and correcting cold weather protection issues.
This completed one sample for a site specific weather related condition.
- December 1-2 with nightly outside temperatures below freezing
b. Findings
No findings were identified.
1R04 Equipment Alignment
.1 Partial Equipment Walkdowns
a. Inspection Scope
The inspectors performed walkdowns of the critical portions of the selected trains to
verify correct system alignment. The inspectors reviewed plant documents to determine
the correct system and power alignments, and the required positions of select valves
and breakers. The inspectors verified that the licensee had properly identified and
resolved equipment alignment problems that could cause initiating events or impact
mitigating system availability. The inspectors verified the following three partial system
alignments during system walkdowns using the listed documents:
- Emergency diesel generator EGDG-1B, raw water pump RWP-2B, and service water
pump SWP-1B trains using operating procedure OP-707, Operation of the ES
Emergency Diesel Generators, and OP-408, Nuclear Services Cooling System, while
the A train systems (EGDG, RW and SW) were out of service to support planned
maintenance
- Emergency diesel generator EGDG-1A and the alternate AC emergency diesel
generator EGDG-1C, using OP-707 and OP-707C, Operation of The Alternate AC
Diesel Generator, while EGDG-1B was out of service for testing
- B train decay heat (DH) and the decay heat closed cycle cooling (DC) systems using
OP- 404, Decay Heat Removal System; and the 4160/480 Volt switch gear rooms,
using OP-703, Plant Distribution System, while the A train DH, DC, and emergency
diesel generator were out of service for maintenance
b. Findings
No findings were identified.
.2 Complete Equipment Walkdown
a. Inspection Scope
The inspectors conducted a detailed review of the condition of the emergency feed water
system (turbine driven emergency feed water pump EFP-2 and the diesel driven EFP-3)
and of the makeup system (makeup pumps 1A, 1B and 1C). A review of outstanding
Enclosure
5
maintenance work orders was performed to verify that any deficiencies did not
significantly affect system function. In addition, the inspectors reviewed NCRs to verify
that system problems were being identified and appropriately resolved. The system
health reports (emergency feed water dated October 29, 2010, and makeup system
dated July 16, 2010) and system equipment walkdown summary reports (makeup and
purification dated July 07, 2010, and emergency feed water dated July 7, 2010), were
reviewed to ensure equipment issues identified were properly addressed in the
corrective action program (CAP). The walkdowns to verify system lineup were not
completed due to delays in returning the systems to service. The completion of this
inspection that will verify proper system lineup will be completed prior to unit restart in
2011. Documents reviewed are listed in the Attachment.
b. Findings
No findings were identified.
1R05 Fire Protection
Fire Area Walkdowns
a. Inspection Scope
The inspectors walked down accessible portions of the plant to assess the licensees
implementation of the fire protection program. The inspectors checked that the areas
were free of transient combustible material and other ignition sources. Also, fire
detection and suppression capabilities, fire barriers, and compensatory measures for fire
protection problems were verified. The inspectors checked fire suppression and
detection equipment to determine whether conditions or deficiencies existed which could
impair the function of the equipment. The inspectors selected the areas based on a
review of the licensees probabilistic risk assessment. The inspectors also reviewed the
licensees fire protection program to verify the requirements of Final Safety Analysis
Report (FSAR) Section 9.8, Plant Fire Protection Program, were met. Documents
reviewed are listed in the Attachment. The inspectors toured the following five areas
important to reactor safety:
- Emergency Feed Water Initiation and Control (EFIC) Rooms
- Emergency feed water pump EFP-3 building
- Make up pump (MUP-1A, 1B and 1C) cubicles
- Fire pump building
- Cable spreading room
b. Findings
No findings were identified.
Enclosure
6
1R06 Flood Protection Measures
Internal Flood Protection
a. Inspection Scope
The Inspectors inspected the manholes listed below that are subject to flooding to verify
cables were not submerged in water, cables were intact, and cable support structures
were adequate to perform its function. The inspectors observed four manholes that are
subject to flooding that contain equipment important for the safe operation of the plant.
Documents reviewed are listed in the Attachment.
- Manhole E-1 (Location: hot machine shop; Circuits: circulating water pump (CWP)
power cables (480 VAC) and intake systems control/alarm circuits)
- Manhole E-2 (Location: Southeast berm; Circuits: CWP power cables and intake
systems control/alarm circuits)
- Manhole E-3 (Location: Southwest berm; Circuits: CWP power cables and intake
systems control/alarm circuits)
- Manhole E-7 (Location: Intake; Circuits: CWP power cables and intake systems
control/alarm circuits)
b. Findings
No findings were identified.
1R11 Licensed Operator Requalification
Annual Review of Licensee Requalification Examination Results
a. Inspection Scope
In February 2010, the licensee completed administering the annual requalification
operating tests which were required to be given to all licensed operators in accordance
with 10 CFR 55.59(a)(2). The inspectors performed an in-office review of the overall
pass/fail results of the individual operating tests, as well as the crew simulator operating
tests. These results were compared to the thresholds established in Manual Chapter
609 Appendix I, Operator Requalification Human Performance Significance
Determination Process.
b. Findings
No findings were identified.
Enclosure
7
1R12 Maintenance Effectiveness
a. Inspection Scope
The inspectors reviewed the licensees effectiveness in performing routine maintenance
activities. The review included an assessment of the licensees practices associated
with the identification, scope, and handling of degraded equipment conditions, as well as
common cause failure evaluations and the resolution of historical equipment problems.
For those systems, structures, and components within the scope of the Maintenance
Rule (MR) per 10 CFR 50.65, the inspectors verified that reliability and unavailability
were properly monitored and that 10 CFR 50.65 (a)(1) and (a)(2) classifications were
justified in light of the reviewed degraded equipment condition. Documents reviewed are
listed in the Attachment. The inspectors conducted this inspection for the following two
equipment issues:
- NCR 269400, Emergency Diesel Generators returned to MR (a)(2)
- NCR 425961, SWP-1A vibration in alert range
In addition, the inspectors reviewed the licensees MR (a)(3) Periodic Evaluation
indicated below to verify that the PE had been completed once per refueling cycle. The
licensee has reviewed MR (a)(1) goals, MR (a)(3) performance criteria, preventative
maintenance and effectiveness of corrective action and industry operating experience.
The licensee has made appropriate adjustments as a result of the PE. This constitutes
an addition sample under this inspection procedure.
- AR 396407396407 Maintenance Rule (a)(3) Periodic Assessment dated May 2010
b. Findings
No findings were identified.
1R19 Post Maintenance Testing
a. Inspection Scope
The inspectors witnessed and/or reviewed post-maintenance test procedures and/or test
activities, as appropriate, for selected risk significant systems to verify: (1) testing was
adequate for the maintenance performed, (2) acceptance criteria were clear, and
adequately demonstrated operational readiness consistent with design and licensing
basis documents, (3) test instrumentation had current calibrations, range, and accuracy
consistent with the application, (4) tests were performed as written with applicable
prerequisites satisfied, and (5) equipment was returned to the status required to perform
its safety function. The five post-maintenance tests reviewed are listed below:
- Surveillance procedure SP-344A, RWP-2A, SWP-1A and Valve Surveillance, (SWP-
1A portion only), after performing unplanned maintenance on SWP-1A per work
order (WO) 1833988
Enclosure
8
- SP-354A-EC-71897-TP2, EC Functional Test Procedure for Protective Relay to
Increase EGDG-A Availability While Paralleled for Testing Generator Run, after
installing modification EC-71897. WO 1846129, which implemented changes to EC- 71897 Revision 11 and tested breakers 3205, 3211, 3207, and 3209
- Nuclear assurance procedure NAP-02, Preparation and Control of CR3 Site Specific
Special Processes and Guidelines, Appendix 5, Ultrasonic Shear-Wave Examination
of Socket Welds, after installation of decay heat valve DHV-215, per WO 1849124
- Performance test PT-399, DCV-17, DCV-177, DCV-18, and DCV-178 Stroke Test (A
train only); and preventative maintenance procedure PM-260, Calibration of Decay
Heat Exchanger Closed Cycle Cooling Control Loop after performing maintenance
on DCV-17 per WO 1838581
Cooling System Fan and valve Surveillance, after maintenance affecting service
water valves SWV-43, SWV-151, SWV-152 and SWV-355 per WOs 1136741 and
1567589
b. Findings
No findings were identified.
1R20 Refueling and Outage Activities
Steam Generator Replacement Refueling Outage (RFO16)
a. Inspection Scope
On September 26, 2009, the unit was shut down for a steam generator replacement
refueling outage. NRC integrated inspection reports 05000302/2009005,
05000302/2010002, 05000302/2010003 and 05000302/2010004 documented NRC
outage inspection activities prior to this inspection period. To verify the licensee was
managing fatigue, the inspectors verified that the outage shift schedule allowed for the
minimum days off in accordance with 10 CFR Part 26. In addition, the inspectors
determined that there were no fatigue waiver requests, fatigue self-declarations and
fatigue assessments since this aspect was last reviewed during the 2010 second quarter
inspection period. The inspectors observed and monitored licensee controls over the
refueling outage activities listed below. Additional inspection results for RFO16 will be
documented in next quarters NRC integrated inspection report 05000302/2011002.
Documents reviewed are listed in the Attachment.
- Outage related risk assessment monitoring
- Controls associated with shutdown cooling, reactivity management, electrical power
alignments, containment closure, and spent fuel pool cooling
- Implementation of equipment clearance activities
- Reduced inventory activities
Enclosure
9
- Refueling activities including verification that fuel assemblies were loaded in the
correct reactor core locations
- Reactor mode changes
b. Findings
No findings were identified. During the creation of a temporary opening in the reactor
containment building to support steam generator replacement, the licensee discovered
an internal crack in the concrete containment. The circumstances associated with the
crack in the concrete containment wall were assessed by an NRC special inspection
team. The results of that inspection are documented in NRC special inspection report
1R22 Surveillance Testing
a. Inspection Scope
The inspectors observed and/or reviewed five surveillance tests listed below to verify
that Improved Technical Specification (ITS) surveillance requirements were followed and
that test acceptance criteria were properly specified. The inspectors verified that proper
test conditions were established as specified in the procedures, that no equipment
preconditioning activities occurred, and that acceptance criteria had been met.
Additionally, the inspectors verified that equipment was properly returned to service and
that proper testing was specified and conducted to ensure that the equipment could
perform its intended safety function following maintenance or as part of surveillance
testing.
In-Service Test:
- SP- 340E, DHP-1B, BSP-1B And Valve Surveillance
Surveillance Test:
- SP- 417, Refueling Interval Integrated Plant Response To An Engineered
Safeguards Actuation
- SP-630, MUP/HPI Check Valves Full Flow Test
b. Findings
No findings were identified.
Enclosure
10
2. RADIATION SAFETY (RS)
Cornerstones: Occupational Radiation Safety (OS) and Public Radiation Safety (PS)
2RS1 Radiological Hazard Assessment and Exposure Controls
a. Inspection Scope
Hazard Assessment and Instructions to workers During facility tours, the inspectors
directly observed labeling of radioactive material and postings for radiation areas, high
radiation areas (HRAs), and airborne radioactivity areas established within the
radiologically controlled area (RCA). The inspectors independently measured radiation
dose rates or directly observed conduct of licensee radiation surveys for selected RCA
areas. The inspectors reviewed survey records for several plant areas including surveys
for alpha emitters, hot particles, airborne radioactivity, gamma surveys with a range of
dose rate gradients, and pre-job surveys for selected Unit 3 (U3) refueling outage (RFO)
work activities. The inspectors also discussed with licensee cognizant representatives
changes to plant operations that could contribute to changing radiological conditions
since the last inspection. For selected U3 RFO jobs, the inspectors attended pre-job
briefings and reviewed radiation work permit (RWP) details to assess communication of
radiological control requirements and current radiological conditions to workers.
Selected work activities included decontamination of the deep end of the cavity, transfer
tube cover work, letdown cooler room work, and insulation work under the reactor
vessel.
Hazard Control and Work Practices The inspectors evaluated access barrier
effectiveness for selected U3 Locked HRA and Very HRA locations. Changes to
procedural guidance for Locked HRA and Very HRA controls were discussed with
selected radiation protection (RP) supervisors. Controls and their implementation for
storage of irradiated material within the spent fuel pool (SFP) were reviewed and
discussed in detail with licensee representatives. Established radiological controls
(including airborne controls) were evaluated for selected tasks including work in auxiliary
building HRAs, and radwaste processing and storage areas. In addition, licensee
controls for areas where dose rates could change significantly as a result of plant
shutdown and U3 refueling operations were reviewed and discussed.
Occupational workers adherence to selected RWPs and RP technician (RPT)
proficiency in providing job coverage were evaluated through direct observations and
interviews with licensee staff. Electronic dosimeter (ED) alarm set points and worker
stay times were evaluated against area radiation survey results for selected U3 RFO
activities. ED alarm logs were reviewed and worker response to dose and dose rate
alarms during selected work activities was evaluated. For HRA tasks involving
significant dose rate gradients, the inspectors evaluated the use and placement of whole
body and extremity dosimetry to monitor worker exposure.
Control of Radioactive Material The inspectors observed surveys of material and
personnel being released from the RCA using small article monitor, personnel
contamination monitor, and portal monitor instruments. The inspectors reviewed records
Enclosure
11
for selected release point survey instruments and discussed equipment sensitivity, alarm
setpoints, and release program guidance with licensee staff. The inspectors compared
recent 10 CFR Part 61 results for the dry active waste (DAW) radioactive waste
(radwaste) stream with radionuclides used in calibration sources to evaluate the
appropriateness and accuracy of release survey instrumentation. The inspectors also
reviewed records of leak tests on selected sealed sources and discussed nationally
tracked source transactions with licensee staff.
Problem Identification and Resolution Nuclear Condition Reports (NCRs) associated
with radiological hazard assessment and control were reviewed and assessed. The
inspectors evaluated the licensees ability to identify and resolve the issues in
accordance with procedure CAP-NGGC-0200, Condition Identification and Screening
Process. The inspectors also evaluated the scope of the licensees internal audit
program and reviewed recent assessment results.
RP activities were evaluated against the requirements of Final Safety Analysis Report
(FSAR) Chapters 11 and 12; Improved Technical Specifications (ITS) Sections 5.4 and
5.8; 10 CFR Parts 19 and 20; and approved licensee procedures. Licensee programs
for monitoring materials and personnel released from the RCA were evaluated against
10 CFR Part 20 and IE Circular 81-07, Control of Radioactively Contaminated Material.
Documents reviewed are listed in the Attachment.
The inspectors completed all specified line-items detailed in Inspection Procedure (IP)
71124.01 (sample size of 1)
b. Findings
No findings were identified.
2RS6 Radioactive Gaseous and Liquid Effluent Treatment and Monitoring Systems
a. Inspection Scope
Effluent Monitoring and Radwaste Equipment During inspector walkdowns, accessible
sections of the liquid and gaseous radwaste and effluent systems were assessed for
material condition and conformance with system design diagrams. The inspection
included floor drain tanks, liquid waste system piping, waste gas decay tanks, monitor
tanks, liquid radwaste monitors, plant stack effluent monitors, and associated airborne
effluent sample lines. The inspectors interviewed licensee staff regarding radwaste
equipment configuration and effluent monitor operation.
The inspectors reviewed performance records and calibration results for selected
radiation monitors, flowmeters, and air filtration systems. For effluent radiation monitors
RMA-3 (main plant stack), RMA-4 (reactor building purge exhaust), RMA-6 (fuel
handling building exhaust), RML-2 Plant Liquid Discharge Line (prior to dilution) and
RML-5 (liquid waste), the inspectors walked down the monitors for material condition
and alignment. The last two surveillances on the control room HEPA/Charcoal air
Enclosure
12
treatment systems also were reviewed. The inspectors evaluated out-of-service effluent
radiation monitors and compensatory action data for the period January 2009 - August
2010.
Installed configuration, material condition, operability, and reliability of selected effluent
sampling and monitoring equipment were reviewed against details documented in the
following: 10 CFR Part 20; Regulatory Guide (RG) 1.21, Measuring, Evaluating and
Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials In Liquid
and Gaseous Effluents from Light-Water Cooled Nuclear Power Plants; American
Nuclear Standards Institute (ANSI)-N13.1-1969, Guide to Sampling Airborne Radioactive
Materials in Nuclear Facilities; ITS Section 5; the Offsite Dose Calculation Manual
(ODCM); and FSAR, Chapter 12. Documents reviewed are listed in the Attachment.
Groundwater The inspectors reviewed the sites groundwater sampling and analysis
results and discussed the current trends with Chemistry Department personnel.
Effluent Release Processing and Quality Control Activities The inspectors reviewed
recent liquid and gaseous release permits including pre-release sampling results,
effluent monitor set-points, and resultant doses to the public. The inspectors also
reviewed the 2008 and 2009 annual effluent reports to evaluate reported doses to the
public and to review ODCM changes. The inspectors reviewed daily Quality Control
(QC) data logs and calibration records for instruments used to quantify effluent sample
activity including High Purity Germanium (HPGe) detectors and liquid scintillation
counters. In addition, results of the 2009, and 2010 inter-laboratory cross-check
program were reviewed.
Observed task evolutions, count room activities, and offsite dose results were evaluated
against details and guidance documented in the following: 10 CFR Part 20 and
Appendix I to 10 CFR Part 50; ODCM; RG 1.21; RG 1.109, Calculation of Annual Doses
to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating
Compliance with 10 CFR Part 50 Appendix I; and ITS Section 6. Documents reviewed
are listed in the Attachment.
Problem Identification and Resolution: Selected NCRs associated with effluent release
activities were reviewed and assessed. The inspectors evaluated the licensees ability
to identify, characterize, prioritize, and resolve selected issues in accordance with
procedure CAP-NGGC-0200, Condition Identification and Screening process, Rev.
33.The inspectors also evaluated the scope of the licensees internal audit program and
reviewed recent assessment results. Documents reviewed are listed in the Attachment.
The inspectors completed one specified line-item sample as detailed in IP 71124.06.
b. Findings
No findings were identified.
Enclosure
13
2RS8 Radioactive Solid Waste Processing and Radioactive Material Handling, Storage, and
Transportation
a. Inspection Scope
Radioactive Material Storage. Inspectors reviewed the FSAR, the Process Control
Program (PCP) and recent effluent release report for information on the types, amounts
and processing of radioactive waste disposed. Quality assurance documents in this
area were reviewed.
During facility walkdowns inspectors observed containers of radioactive waste for proper
labeling in accordance with 10 CFR 20.1904 and 10 CFR 20.1905. Inspectors noted the
satisfactory material condition of containers throughout the facilities inside and out.
Postings around stored radioactive materials were in accordance with 10 CFR 20.
Radioactive Waste System Inspectors evaluated the waste disposal systems during
plant walkdowns, escorted and unescorted, and through discussion with cognizant
employees. Accessible components of the liquid and solid waste systems were
observed some of the many areas that were evaluated such as the Yellow room;
laundry/shower sump tank rooms; miscellaneous waste storage tank pump room; and
the RC evaporator valve alley. Processes for transferring radioactive waste into disposal
containers were reviewed by inspectors.
Radioactive waste characterization and shipping The Annual Effluence report for 2009
were reviewed by inspectors. Major waste streams were reviewed for the primary and
secondary resins, reactor coolant filters and DAW. Inspectors evaluated the analysis for
hard-to-detect nuclides, looked at scaling factors, and reviewed the quality assurance
(QA) comparison results between the licensees contracted laboratory results and
outside laboratory results.
Radwaste processing activities and equipment configuration were reviewed for
compliance with the licensees PCP and FSAR, Chapter 11. Waste stream
characterization analyses were reviewed against regulations detailed in 10 CFR Part 20,
10 CFR Part 61, and guidance provided in the Branch Technical Position on Waste
Classification (1983). Documents reviewed are listed in the Attachment.
Transportation program implementation was reviewed against regulations detailed in 10
CFR Part 20, 10 CFR Part 71, 49 CFR Parts 172-178, as well as the guidance provided
in NUREG-1608. Training activities were assessed against 49 CFR Part 172 Subpart H.
Documents reviewed are listed in the Attachment.
Problem Identification and Resolution The inspectors reviewed selected NCRs and
audits associated with radioactive solid waste processing and radioactive material
handling, storage and transportation. The inspectors evaluated the licensees ability to
identify, characterize, prioritize, and resolve the identified issues in accordance with
procedure CAP-NGGC-0200, Corrective Action Program, Rev. 33.
Enclosure
14
The inspectors completed one sample as detailed by inspection procedure 71124.08.
b. Findings
No findings were identified.
4. OTHER ACTIVITIES
4OA1 Performance Indicator (PI) Verification
.1 Reactor Safety
a. Inspection Scope
The inspectors checked the mitigating system performance indicators (MSPI) listed
below to verify the accuracy of the PI data reported. Performance indicator data
submitted from October 2009 through September 2010 was compared for consistency to
data obtained through review of monthly operating reports, nuclear condition reports,
and control room logs. The inspections were conducted in accordance with NRC
Inspection Procedure 71151, Performance Indicator Verification. The applicable
planning standard, Nuclear Energy Institute (NEI) 99-02, Revision 6, Regulatory
Assessment Performance Indicator Guidelines, and the licensees calculation P06-0002,
CR3 MSPI Basis Document for the CR3 Nuclear Plant, were used to check the reporting
for each data element. The inspectors discussed the PI data with the licensee personnel
associated with performance indicator data collection and evaluation.
- Emergency AC power
- Residual heat removal/decay heat system
- Heat removal system
- High pressure injection system
- Cooling water system
b. Findings
No findings were identified.
.2 Radiation Safety
a. Inspection Scope
The inspectors sampled licensee data for the PIs listed below. To verify the accuracy of
the PI data reported during the period reviewed, PI definitions and guidance contained in
NEI 99-02, Regulatory Assessment Indicator Guideline, Rev. 6, were used to verify the
basis for each data element.
Enclosure
15
Occupational Radiation Safety (ORS) Cornerstone
The inspectors reviewed Performance Indicator (PI) data collected from October 1, 2009
through September 30, 2010, for the Occupational Exposure Control Effectiveness PI.
For the reviewed period, the inspectors assessed CAP records to determine whether
HRA, VHRA, or unplanned exposures, resulting in ITS or 10 CFR 20 non-conformances,
had occurred during the review period. In addition, the inspectors reviewed selected
personnel contamination event data, internal dose assessment results, and ED alarms
for cumulative doses and/or dose rates exceeding established set-points. The reviewed
data were assessed against guidance contained in NEI 99-02, "Regulatory Assessment
Indicator Guideline," Rev. 6. Documents reviewed are listed in the Attachment.
Public Radiation Safety (PS) Cornerstone
The inspectors reviewed the Radiological Effluent Technical Specification/Offsite Dose
Calculation Manual Radiological Effluent Occurrences PI results from October 1, 2009
through September 30, 2010. The inspectors reviewed CRs, effluent dose data, and
licensee procedural guidance for classifying and reporting PI events. The inspectors
also interviewed licensee personnel responsible for collecting and reporting the PI data.
Documents reviewed are listed in the Attachment.
The inspectors completed 2 of the required 2 samples for IP 71151.
b. Findings
No findings were identified.
4OA2 Identification and Resolution of Problems
.1 Daily Review
a. Inspection Scope
As required by Inspection Procedure 71152, Identification and Resolution of Problems,
and in order to help identify equipment failures or specific human performance issues for
follow-up, the inspectors performed a daily screening of items entered into the licensees
corrective action program (CAP). This review was accomplished by attending daily plant
status meetings, interviewing plant operators and applicable system engineers, and
accessing the licensees computerized database.
b. Findings
No findings were identified.
Enclosure
16
.2 Annual Trend Review
a. Inspection Scope
As required by Inspection Procedure 71152, Identification and Resolution of Problems,
the inspectors performed a review of the licensees CAP and associated documents to
identify trends that could indicate the existence of a more significant safety issue. The
inspectors review was focused on repetitive equipment issues, but also considered the
results of daily inspector CAP item screening discussed in Section 4OA2.1, plant status
reviews, plant tours, and licensee trending efforts. The inspectors review nominally
considered the six month period of July 2010 through December 2010. The review also
included issues documented in the licensees Plant Health Committee Site Focus List
dated November 2010, and various 3rd quarter 2010 departmental CAP Rollup & Trend
Analysis reports, nuclear oversite reports and maintenance rule (MR) reports.
Corrective actions associated with a sample of the issues identified in the licensees
corrective action program were reviewed for adequacy.
b. Findings and Observations
No findings were identified. The inspectors evaluated the licensees trend methodology
and determined that the licensee had performed a detailed review.
.3 Annual Sample Review
a. Inspection Scope
The inspectors selected NCR 431407 for a detailed review and discussion with the
licensee. This NCR was classified as significance level one requiring an apparent cause
evaluation. The NCR investigated an issue where a recently implemented engineering
change (EC) had been installed on the A emergency diesel generator (EGDG). As
installed, the EGDG would not have been able to load onto the A train emergency
service bus (bus) under three separate conditions. The inspectors checked that the
issues had been completely and accurately identified in the licensees corrective action
program; safety concerns were properly classified and prioritized for resolution; apparent
cause determination was sufficiently thorough; and appropriate corrective actions were
initiated. The inspectors also evaluated the NCR using the requirements of the
licensees CAP as delineated in corrective action procedure CAP-NGGC-0200,
Condition Identification And Screening Process.
b. Findings and Observations
On November 3, 2010, the licensee was performing surveillance test procedure SP-902,
4160 ES Bus A Under Voltage Trip Test And Auxiliary Relay Calibration, and found
that breaker 3211 could not be opened when using the control room switch. The Unit
was in a no-mode condition. The reactor vessel contained no fuel and all fuel was
stored in the spent fuel pool. The unit had been in this mode since October 9, 2009, as
a result of the extended refueling outage. The surveillance procedure isolates the bus
from the off-site power transformer (OPT) by opening breaker 3211 using a manual
Enclosure
17
control switch from the control room. During this step, breaker 3211 would not open and
remained in the closed position. The bus could not be isolated from the OPT using the
control room switch. The licensee backed out of the surveillance and entered the issue
into the CAP. The licensees investigation found that the design of a recently installed
EC had incorrectly removed a control wire for breaker 3211. The licensees apparent
cause evaluation identified that the engineering analysis for the EC lacked appropriate
depth and detail. Neither the responsible engineer nor the independent verifier had
adequately analyzed the proposed design change, which resulted in the failure to
complete an electrical connection required for proper operation of breaker 3211.
Corrective actions to address this issue were comprehensive and included training and
an engineering stand down to review this issue. Additionally, the licensee reviewed
other recently designed and installed ECs to verify adequate design and analyses of
correct depth and detail. No additional deficiencies were identified. A licensee identified
violation of design control was assessed by the inspectors and is documented in Section
4OA7.
4 Annual Sample Review - Operator Work Around
a. Inspection Scope
The inspectors reviewed the operator workaround program to verify the licensee was
identifying workarounds at an appropriate threshold and entering them into the corrective
action program. One operator workaround associated with the control complex chiller
system reliability (NCR 379560) was identified that will be resolved during the next
refueling outage. The inspectors determined that compensatory actions in place are
adequate to address the issue.
b. Findings
No findings were identified.
4OA5 Other Activities
.1 Quarterly Resident Inspector Observations of Security Personnel Activities
a. Inspection Scope
During the inspection period, the inspectors conducted observations of security force
personnel and activities to ensure that the activities were consistent with licensee
security procedures and regulatory requirements relating to nuclear plant security.
These observations took place during normal and off-normal plant working hours. These
quarterly resident inspector observations of security force personnel and activities did
not constitute any additional inspection samples. Rather, they were considered an
integral part of the inspectors normal plant status reviews and inspection activities.
b. Findings
No findings were identified.
Enclosure
18
.2 Steam Generator Replacement Project (SGRP) and Containment Wall Repair (IP
50001)
a. Inspection Scope
The inspectors conducted a review of the licensees Phase 4 concrete placement and
Phase 5 retensioning activities for the repair of the containment wall delamination and
reinstallation of the containment wall opening that was created during the SGRP in the
last quarter of 2009.
Rebar and Formwork Installation
The inspectors examined the rebar installation on Elevation 206 + 5 that was prepared
for concrete pour to ensure that the licensee had measured the reinforcing steel size,
spacing, lap splice length, and concrete minimum protection coverage. The inspectors
determined whether the licensee performed inspections on installation, testing, and
testing frequencies of swaged mechanical splices in accordance with the requirements
of the design drawings, the American Concrete Institute (ACI) Codes, and the American
Society of Mechanical Engineers (ASME) Code Section III Division 2, Concrete
Containment. The inspectors also examined the formwork installation and tendon
sleeve condition.
Concrete Pour
The inspectors reviewed the concrete pre-placement inspection checklist, including
cleaning and debris removal prior to the concrete pour. The inspectors observed
concrete placement activities on Elevation 206 + 5 to verify that activities pertaining to
concrete delivery time, flow distance, layer thickness and concrete consolidation or
vibration conformed to industry standards established by the ACI Codes. Concrete
batch tickets were examined to verify the material and quantity of each component for
concrete mix, truck revolution limit, concrete placement time limit, and water amount
added to the mix. The inspectors observed that concrete placement activities were
continuously monitored by the licensee and contractors quality control personnel and
engineers. The inspectors witnessed in-process testing and reviewed the results for
slump, air content, temperature, unit weight, and molding of the concrete cylinders for
compressive strength testing, and witnessed sample points and truck loads to verify that
concrete samples for the field testing and cylinders for the laboratory testing were
obtained at the point of placement (end of chute line) and the middle portion of the truck
loads. The inspectors reviewed concrete being poured into cylinders to determine
whether it was molded in accordance with applicable American Society for Testing and
Materials (ASTM) requirements of ASTM C 172, Standard Method of Sampling Freshly
Mixed Concrete, and to determine whether appropriate concrete field testing was
performed by Quality Control (QC) inspectors.
The inspectors reviewed or examined the licensee activities to verify that the activities
met the ACI code requirements, licensee procedures, and the industry standards. The
inspectors examined the batch plant for its certification and the preparation of the
concrete pour.
Enclosure
19
Containment Retensioning and Testing
The inspectors reviewed the containment retensioning plan, testing plan, and schedule.
The inspectors interviewed licensee personnel and reviewed documents related to the
retensioning and testing plans. The licensee was conducting a detailed analysis to
develop a tendon retensioning sequence that would minimize the possibility of causing
new cracks or delaminations in the containment during the retensioning process. The
licensee is scheduled to perform a Structural Integrity Test (SIT) after final retensioning
in order to test the structural integrity of the containment. Following the SIT, the licensee
is scheduled to perform an Integrated Leak Rate Test (ILRT) on the containment. The
inspectors review included the licensees plans for post-maintenance testing after
restart.
Document Review
The inspectors reviewed the engineering changes (ECs), specifications, drawings, work
packages, nuclear condition reports (NCRs), concrete compressive testing results, and
documents related to the concrete construction activities. The inspectors reviewed EC 75220, Reactor Building Delamination Repair Phase 4 - Concrete Placement, Revision
21 and EC 75221, Reactor Building Delamination Repair Phase 5 - Retensioning,
Revision 0. The inspectors reviewed SGT Work Packages (WP) 3-3732A, B, C, and D
Restoration of Containment Concrete Wall. The reviews or observations were
conducted in order to verify that the licensee performed activities in accordance with the
approved documents.
The inspectors reviewed records to verify that they met the licensee administrative
control procedures, Quality Control standard, Quality Assurance Program requirements,
and applicable industrial design and construction standards.
b. Findings
No findings were identified.
.3 Annual Sample Review
a. Inspection Scope
The inspectors selected NCR 431407 for a detailed review and discussion with the
licensee. This NCR was classified as significance level one requiring an apparent cause
evaluation. The NCR investigated an issue where a recently implemented engineering
change (EC) had been installed on the A emergency diesel generator (EGDG). As
installed, the EGDG would not have been able to load onto the A train emergency
service bus (bus) under three separate conditions. The inspectors verified that the
issues had been completely and accurately identified in the licensees corrective action
program, safety concerns were properly classified and prioritized for resolution, apparent
cause determination was sufficiently thorough, and appropriate corrective actions were
initiated. The inspectors also evaluated the NCR using the requirements of the
Enclosure
20
licensees CAP as delineated in corrective action procedure CAP-NGGC-0200,
Condition Identification And Screening Process.
b. Findings and Observations
On November 3, 2010, the licensee was performing surveillance test procedure SP-902,
4160 ES Bus A Under Voltage Trip Test And Auxiliary Relay Calibration, and found
that breaker 3211 could not be opened when using the control room switch. The Unit
was in a no-mode condition. The reactor vessel contained no fuel and all fuel was
stored in the spent fuel pool. The unit had been in this mode since October 9, 2009, as
a result of the extended refueling outage. The surveillance procedure isolates the bus
from the off-site power transformer (OPT) by opening breaker 3211 using a manual
control switch from the control room. During this step, breaker 3211 would not open and
remained in the closed position. The bus could not be isolated from the OPT using the
control room switch. The licensee backed out of the surveillance and entered the issue
into the CAP. The licensees investigation found that the design of a recently installed
EC had incorrectly removed a control wire for breaker 3211. The licensees apparent
cause evaluation identified that the engineering analysis for the EC lacked appropriate
depth and detail. Neither the responsible engineer nor the independent verifier had
adequately analyzed the proposed design change, which resulted in the failure to
complete an electrical connection required for proper operation of breaker 3211.
Corrective actions to address this issue were comprehensive and included training and
an engineering stand down to review this issue. Additionally, the licensee reviewed
other recently designed and installed ECs to verify adequate design and analyses of
correct depth and detail. No additional deficiencies were identified. A licensee identified
violation of design control was assessed by the inspectors and is documented in Section
4OA7.
.4 (Closed) NRC Temporary Instruction (TI) 2515/179, Verification of Licensee Responses
to NRC Requirement for Inventories of Materials Tracked in the National Source
Tracking System Pursuant to Title 10, Code of Federal Regulations, Part 20.2207 (10
CFR 20.2207)
a. Inspection Scope
The inspectors performed the TI concurrent with IP 71124.01 Radiation Hazard Analysis.
The inspectors reviewed the licensees source inventory records and identified the
sources that met the criteria for reporting to the NSTS. The inspectors visually identified
the sources contained in various calibration systems and verified the presence of the
source by direct radiation measurement using a calibrated portable radiation detection
survey instrument. The inspectors reviewed the physical condition of the irradiation
device. The inspectors reviewed the licensees procedures for source receipt,
maintenance, transfer, reporting and disposal. The inspectors reviewed documentation
that was used to report the sources to the NSTS. Documents reviewed are listed in the
Attachment.
Enclosure
21
b. Findings and Observations
No findings were identified. The inspectors performed the TI concurrent with IP
71124.01 Radiation Hazard Analysis. The inspectors reviewed the licensees source
inventory records and identified the sources that met the criteria for reporting to the
NSTS. The inspectors visually identified the sources contained in various calibration
systems and verified the presence of the source by direct radiation measurement using
a calibrated portable radiation detection survey instrument. The inspectors reviewed the
physical condition of the irradiation device. The inspectors reviewed the licensees
procedures for source receipt, maintenance, transfer, reporting and disposal. The
inspectors reviewed documentation that was used to report the sources to the NSTS
Documents reviewed are listed in the Attachment.
.5 Operator Licensing Training and Qualification Effectiveness Inspection
a. Inspection Scope
The inspectors reviewed associated documents in preparation for this inspection.
During the week of October 25 - 27, 2010, the inspectors reviewed documentation,
interviewed licensee personnel, and observed the administration of training associated
with the licensees operator requalification program and the Just In Time training
associated with the licensees startup preparations following an extended refueling
outage. The inspectors conducted the inspection under the guidance of IP 41500,
Training and Qualification Effectiveness Inspection. The inspectors evaluated that the
licensee had performed, or had scheduled to be performed, training as specified in a
letter from M. Widmann to J. Franke dated March 8, 2010. The inspectors directly
observed three unevaluated simulator scenarios for training, including the operating
crews self-critique; and reviewed the evaluated simulator scenario that was to be
administered to all licensed operators for this training cycle. The inspectors directly
observed classroom training that was given on the integrated plant start-up procedure,
including a presentation from chemistry personnel on some of the off-normal chemistry
concerns that were anticipated during the plant startup. The inspectors reviewed
documentation to include licensee self-assessment reports, watchstanding records for
proficiency, training attendance records, overall training plans and schedules, individual
training lesson plans, and documentation associated with evaluated simulator scenarios.
Documents reviewed are listed in the Attachment.
b. Findings
No findings were identified.
.6 (Closed) NRC Temporary Instruction 2515/172, Reactor Coolant System Dissimilar
Metal Butt Welds
a. Inspection Scope
The inspectors conducted a review of the licensees activities regarding licensee
dissimilar metal butt weld (DMBW) mitigation and inspection implemented in accordance
Enclosure
22
with the industry self imposed mandatory requirements of Materials Reliability Program
(MRP) 139, Primary System Piping Butt Weld Inspection and Evaluation Guidelines.
Temporary Instruction (TI) 2515/172, Reactor Coolant System Dissimilar Metal Butt
Welds, Revision 1 was issued May 27, 2010, to support the evaluation of the licensees
implementation of MRP-139.
On December 8, 2010, the inspectors performed a review in accordance with TI
2515/172, Revision 1 as described in the Observation Section below:
b. Observations
The licensee has met the MRP-139 deadlines for baseline examinations of all welds
scoped into the MRP-139 program. TI 2515/172, Revision 1 is considered closed. In
accordance with requirements of TI 2515/172, Revision 1, the inspectors evaluated the
following areas:
(1) Implementation of the MRP-139 Baseline Inspections
This portion of the TI was not inspected during the period of this inspection report, but
was previously covered in NRC Inspection Report 05000302/2008002.
(2) Volumetric Examinations
This portion of the TI was not inspected during the period of this inspection report, but
was previously covered in NRC Inspection Report 05000302/2009005.
(3) Weld Overlays
This portion of the TI was not inspected during the period of this inspection report, but
was previously covered in NRC Inspection Report 05000302/2008002.
(4) Mechanical Stress Improvement (SI)
There were no stress improvement activities performed or planned by this licensee to
comply with their MRP-139 commitments.
(5) Application of Weld Cladding and Inlays
There were no weld cladding nor inlay activities performed or planned by this licensee to
comply with their MRP-139 commitments.
(6) Inservice Inspection Program
This portion of the TI was not inspected during the period of this inspection report, but
was previously covered in NRC Inspection Report 05000302/2008005.
c. Findings
No findings were identified.
Enclosure
23
4OA6 Meetings, Including Exit
Exit Meeting Summary
On January 10, 2011, the resident inspectors presented the inspection results to Mr. J.
Franke, Site Vice President, and other members of licensee management. The
inspectors confirmed that proprietary information was not provided or examined during
the inspection.
4OA7 Licensee Identified Violations
The following issue of very low safety significance (Green) was identified by the licensee
and was a violation of NRC requirements. This issue met the criteria of Section 2.3.2 of
the NRC Enforcement Policy, NUREG-1600, for being dispositioned as a Non-Cited
Violation.
10 CFR 50 Appendix B, Criterion III, Design Control, requires, in part, that measures
shall be established to assure that applicable regulatory requirements and design basis
for those structures, systems, and components are correctly translated into
specifications, drawings, procedures and instructions. Engineering corporate
procedures EGR-NGGC-0011, Engineering Rigor; and EGR-NGGC-0155, Specifying
Electrical / I&C Modification Related Tests, implement those requirements. Contrary to
the above, the licensee failed to translate the design basis into drawings and procedures
when performing design modification EC 71897. This resulted in an electrical circuit
error in the A EDG breaker logic circuitry. The inadequate EC removed a switchgear
internal control wire that supplied DC control power to the following: 1) OPT differential
lockout relay to trip breaker 3211, 2) MCB control switch open contacts to trip breaker
3211, and 3) emergency safety A-bus under-voltage trip circuit to trip breaker 3211. As
a result of breaker 3211 not being able to trip under any of these three signals, the A
EDG would not have been able to meet the logic required to load onto the safety bus
when required. The licensee determined that engineering personnel did not have an
adequate understanding of assessing the correct engineering depth and detail involved
in designing and implementing the EC. The process deficiency of failing to provide
adequate depth and detail on the EC is more than minor because, if left uncorrected,
would have the potential to lead to a more significant safety concern. The finding was
determined to be of very low safety significance (Green) because there were no diesel
operability requirements during the time the inadequate EC had been installed.
Additionally, the inadequate EC was identified and corrected by the licensee prior to the
emergency generator being required by plant technical specifications to be available to
support a change in mode. This issue was documented in the licensees corrective
action program as NCR 431407. Additional information regarding this issue can be
found in Section 4OA2.3.
ATTACHMENT: SUPPLEMENTAL INFORMATION
Enclosure
KEY POINTS OF CONTACT
Licensee personnel:
B. Akins, Superintendent, Radiation Protection
M. Bishara, SGR Design Engineering Manager
S. Cahill, Manager, Engineering
J. Cravens, SGR Welding Engineer
F. Dola, Nuclear Oversight Superintendent
P. Dixon, Manager Training
D. Douglas Manager, Maintenance
P. Fagan, Repair Design and Construction Engineering Supervisor
J. Franke, Vice President, Crystal River Nuclear Plant
R. Griffith, SGR Task Manager
K. Henshaw, SGR Rigging Supervisor
D. Herrin, Licensing Engineer
J. Holt, Plant General Manager
J. Huegel, Manager, Nuclear Oversite
D. Jopling, SGR Civil Structural Supervisor
B. Kelley, RT Level III
D. Mayes, SGR Welding Engineer
W. Nielsen, SGR QC Supervisor
C. Poliseno, Supervisor, Emergency Preparedness
S. Powell, SGR Licensing engineer
J. Terry, SGR Project Manager
R. Vessley, SGR QC Supervisor
D. Westcott, Supervisor, Licensing
I. Wilson, Manager Outage and Scheduling
B. Wunderly, Manager, Operations
NRC personnel:
D. Rich, Chief, Branch 3, Division of Reactor Projects
LIST OF ITEMS OPENED, CLOSED AND DISCUSSED
Closed
05000302/2515/179 TI Verification of Licensee Responses to NRC
Requirement for Inventories of Materials
Tracked in the National Source Tracking
System Pursuant to Title 10, Code of
Federal Regulations, Part 20.2207 (10 CFR
20.2207), Revision 1 (Section 4OA5.4)
05000302/2515/172 TI Reactor Coolant System Dissimilar Metal
Butt Welds Revision 1 (Section 40A5.6)
Attachment
2
Discussed
05000302/2515/177 TI Managing Gas Accumulation in Emergency
Core Cooling, Decay Heat Removal, and
Containment Spray Systems (NRC Generic
Letter (GL) 2008-01) Revision 1 (Section
4OA5.3)
LIST OF DOCUMENTS REVIEWED
Section 1R04: Equipment Alignment
Nuclear Condition Reports (NCRs)
NCR 266531, EFGV-1 Governor oil out of sight high
NCR 286943, EFV-11 Closed without operator action
NCR 287855, EFP-3 Time delay relay setting different from DBD
NCR 301727, Temporary change from not completed within 14 days
NCR 304139, Untimely update of EF maintenance rule event log by system Engineer
NCR 304742, SP-146 Temp Change
NCR 305239, EFP-1 Breaker 52/H deficiency following maintenance
NCR 310438, Emergency feed water control valve stroke time repeatability
NCR 326879, EFV-148, Actuator stripped internally
NCR 331000, Blank flange found removed from EFT-2 building penetration
NCR 354537, SP-146 EFIC D channel would not go to maintenance bypass
NCR 363244, EFIC CH B blown fuse
NCR 371967, EFT-2 oxygen values not expected due to hydrazine levels
NCR 378992, EFP-2 found rotating during standby (on 1&2 aux steam)
NCR 266361, RECL-256 does not agree with RECL-267 within 14.1#
NCR 276666, Failed optical isolator during performance of SP-146
NCR 319008, DFT-4 particulate increasing trend near alert limit
NCR 325123, DFP-2 vibration data per SP-349B has low margin to IST alert
NCR 328103, Surface corrosion noted on EF piping in EFP-3 building
NCR 387621, EF buried piping G-scan inspection results
NCR 296860, PM-152 soil to pipe potentials found out of specification
NCR 295834, EF piping under west berm requires evaluation
NCR 294773, EF piping cathodic protection not performed annually
NCR 427428, MUV-244 significant packing leak
NCR 424905, MUP-1B motor phase resistance inbalance
NCR 400263, Piping not fully qualified
NCR 369532, NRC GL 2008-01 implementation challenges
NCR 358946, MUV-36 check valve disc separation
Section 1R05: Fire Protection
Procedures
AI-2205A, Pre Fire Plan - Control Complex
AI-2205B, Pre Fire Plan - Turbine Building
AI-2205C, Pre Fire Plan - Auxiliary Building
Attachment
3
Al -2205F, Pre Fire Plan - Miscellaneous buildings and Components
SP-804, Surveillance of Plant Fire Brigade Equipment
Section 1R06: Flooding
Model Work Order 614384, Site Manholes and Handhole Inspections
Implementation Work Order 1646872, Site Manholes and Handhole Inspections
Section 1R12: Maintenance Effectiveness
System engineering report SE10-0040, EG to be re-classified as (a)(2)
NCR 265002, Jacket coolant leaks identified during SP-354B
NCR 269400, EGDG Jacket adapter minor leak
WO 01600340, EGDG-1A/B/C EC-73336 for Dresser coupling restraint devices
WO 01357712, Replace Dresser coupling gaskets on EGDG-1A And EGDG-1B
NCR 400460, Minor errors found during periodic assessment
NCR 400462, Maintenance Rule evaluations corrections
AR 262703262703 Maintenance Rule Program (a)(3) Periodic Assessment dated May 2008
Licensee administrative procedure ADM-NGGC-0101, Maintenance Rule Program
WO 1848531, BSP-1B Coupling inspection and lubrication
Section 1R20: Refueling and Outage Activities
Procedures
AI-504, Guidelines for Cold Shutdown and Refueling
CP-341, Containment Penetration Control
FP-410, Reactor Vessel Closure Head Installation
FP-203, Offloading And Refueling Operations
OP-301A, Refueling Outage RCS Drain and Fill Operations
OP-421A, Operation Of The Reactor Building Polar Crane RCCR-1
SP-440, Unit Startup Surveillance Plan
WCP-102, Outage Risk Assessment
Nuclear Condition Reports
NCR 247148, Industry initiatives on heavy loads
NCR 314049, Crystal River 3 Outage Risk Assessment for R16
Calculations
Nureg - 0612 Nine-Month Report Control of Heavy Loads at Nuclear Plants, Crystal River Unit
3, Appendix F, Analysis of the Effect of Reactor Vessel Head Drop on the Reactor Vessel,
October 1983
Section 2RS1: Radiological Hazard Assessment and Exposure Controls
Procedures, Guidance Documents, and Manuals
CAP-NGGC-0200, Condition Identification and Screening Process, Rev. 33
CAP-NGGC-0201, Self-Assessment/Benchmark Programs, Rev. 14
CAP-NGGC-0205, Condition Evaluation and Corrective Action Process, Rev. 12
Attachment
4
CP-123, Restrained Components and Key Control, Rev, 61
DOS-NGGC-0002, Dosimetry Issuance, Rev. 27
FP-605, Spent Fuel Pools Controls and Accountability, Rev. 4
HPP-112, Hard to Detect Radionuclides Analyses, Rev. 2
HPP-202A, Supplemental Instructions to HPS-NGGC-0003: Radiological Surveys and
Inspections, Rev. 35
HPP-215, Health Physics Source Receipt and Control, Rev. 14
HPP-216, Diving Operations in Radiological Environments, Rev. 9
HPP-221, High Radiation Area, Locked High Radiation Area, and Very High Radiation Area
Controls, Rev. 12
HPS-NGGC-0001, Radioactive Material Receipt and Shipping Procedure, Rev. 30
HPS-NGGC-0003, Radiological Posting, Labeling and Surveys, Rev. 15
HPS-NGGC-0013, Personnel Contamination Monitoring, Decontamination, and Reporting,
Rev. 12
HPS-NGGC-0014, Radiation Work Permits, Rev. 8
HPS-NGGC-0016, Access Control, Rev. 6
HPS-NGGC-0023, Remote Radiological Monitoring, Rev. 4
HPS-NGGC-0024, Alpha Monitoring Guidelines, Rev. 2
SEC-NGGC-2162, Keys, Locks, and Combinations, Rev. 3
SS-201, Security Force Personnel General Orders, Duties, and Responsibilities, Rev. 60
Records and Data
2010 DAW Smears, Dated 09/08/10
CP-123, Restrained Components and Key Control, Rev, 61, Enclosure 5, Key Control
Log, Selected Logbook Entries
HPP-221, High Radiation Area, Locked High Radiation Area, and Very High Radiation Area
Controls, Rev. 12, Enclosures 1, HP Key Control Log; and 2, LHRA/SRC Authorization Key
Control Log, Selected Logbook Entries
Health Physics Survey Record (HPSR), Survey No. 10-01-0023, Routine HPP-202Z02,
HP Sources, Dated 01/07/10
HPSR, Survey No. 10-01-0210, Routine HPP-202W14, Survey of Source Storage Area,
Dated 10/15/10
HPSR Survey No. 10-10-0414, 95 Reactor Building Inside and Outside of Letdown
Cooler Room, Dated 10/28/10
HPSR Survey No. 10-11-0444, Under Vessel, Dated 11/30/10
HPSR Survey No. 10-12-0062, Old Letdown Cooler Room, Dated 12/06/10
HPSR Survey No. 10-12-0071, Reactor Building Lower Cavity, Dated 12/06/10
HPSR Survey No. 10-12-0082, Lower Reactor Cavity, Dated 12/07/10
NRC Form 748, National Source Tracking Transaction Report, Crystal River 3 Nuclear Power
Plant, License. No. DPR-72, Docket No. 50-302, Dated 01/05/2010
Personnel Contamination Log, RFO 16
Radiation Work Permit (RWP) Number (No.) 4704, Non-SGR Rad Waste Activities (High
Risk)
RWP No. 4711, Non-SGR Reactor Head/Cavity Activities (High Risk)
RWP No. 4732, Non-SGR Maintenance Activities (High Risk)
RWP No. 4744, Non-SGR Insulation Activities (High Risk)
SEC-NGGC-2162, Keys, Locks, and Combinations, Rev. 11/30/10, Attachment 1,
Attachment
5
Security Shift Key Control and Inventory, Selected Logbook Entries
Spent Fuel Pool Storage SFPA, Dated 10/20/10
Corrective Action Program (CAP) Documents
Action Request 00436795, Red plastic bags used in the RCA, Dated 12/07/10
Assessment No. 310235, Radiation Protection Refuel 16 Readiness, Undated
Assessment No. 382005, Quick Hit Self-Assessment Report, Electronic Dosimetry Set
Points, Undated
Section: 2RS6 Radioactive Gases and Liquid Effluent Treatment and Monitoring Systems
Procedures, Guidance Documents, and Manuals
CP-161, Radiological Environmental Monitoring Program, Rev. 6
SP-187, AHFL-2A/2B/2C/2D (Auxiliary Building) In-Place Filter Testing, Rev. 29
SP-731A, Auxiliary Building Ventilation Continuous Release, Rev.11
SP-731B, Reactor Building Purge Batch Release and Batch to Continuous Release, Rev. 21
SP-731C, Reactor Building Ventilation Continuous Release, Rev. 14
SP-731F, WDT-1A/WDT-1B/WDT-1C Release, Rev.10
SP-731E, Reactor Building Atmospheric Release after Integrated Leak Rate Testing, Rev. 9
SP-736A, WDT-10A Release to the Discharge Canal, Rev. 11
SP-736B, WDT-10B Release to the Discharge Canal, Rev. 12
SP-736C, WDT-11A Release to the Discharge Canal, Rev. 8
SP-736D, WDT-11B Release to the Discharge Canal, Rev. 8
SP-736E, WDT-11A and WDT-11B Combined Release to the Discharge Canal, Rev. 11
SP-736F, SDT-1 / Turbine Building Sump / Condensate Release To The Settling Ponds, Rev.13
SP-736G, SDT-1 Release to the Discharge Canal, Rev. 13
SP-736I, Condensate Release to the Discharge Canal, Rev. 10
Crystal River Unit 3 Off-Site Dose Calculation Manual, Rev. 32
Records and Data Reviewed
Crystal River Unit 3 -2008 Radioactive Effluent Release Report, April 21, 2009
Crystal River Unit 3 -2009 Radioactive Effluent Release Report, April 29, 2010
Release Permit 100138.006, 838.L SDT-1 Release to the Discharge Canal
Release Permit 100137.001.721.L WDT-10A Release to the Discharge Canal
Release Permit 100041.020.571.G Auxiliary Building Ventilation Continuous Release
Release Permit 100044. 018.063. G Reactor Containment Building Purge Batch Release
Results of Radiochemistry Cross Check Program 1st Quarter 2009
Results of Radiochemistry Cross Check Program 1st Quarter 2010
Results of Radiochemistry Cross Check Program 2nd Quarter 2010
Results of Radiochemistry Cross Check Program 3rd Quarter 2010
Surveillance: SP-186 AHFL-4A (Control Room) In-Place Filter Testing, 9/4/2009
Surveillance: SP-186 AHFL-4B (Control Room) In-Place Filter Testing, 9/4/2009
Ground Water Tritium Sampling Results from 2/27/2007 to 9/8/2010
CAP-NGGC-0200, Condition Identification and Screening Process, Rev.33
Crystal River Unit 3 UFSAR Chapter 11, Radioactive Waste & Radiation Protection, Rev. 32
Crystal River 3 10 CFR61 Waste Stream Analysis Results, 9/8/2010
Attachment
6
CAP Documents
NCR 411245, Nitrogen Line outside the RCA Spilled a Small Amount of Contaminated Water
(<1 gallon).
Section 2RS8: Radioactive Solid Waste Processing and Radioactive Material Handling,
Storage, and Transportation
Procedures, Manuals, and Guidance Documents
HPS-NGGC-0001, Radioactive Material and Shipping Procedure, Revision 30
HPP-240, Sampling for Part 61 Waste Stream Analysis, Rev. 0
HPS-NGGC-0002, Vendor Cask Utilization Procedure, Revision 17
Miscellaneous General Manual (MGM), Volume 1, Part 1 PCP, Process Control Program,
Revision 6
WP-105, Radioactive Material Shipping Forms, Revision 2
Records and Data Reviewed
2009 Radioactive Waste Shipment Log
2010 Radioactive Waste Shipment Log
Quick Hit Self-Assessment Report: Storage and Control of Licensed Material, 03/16-26/2009
Shipment Number (No.)09-142, Radioactive material Type A package, 7, UN2915, 11/13/2009
Shipment No.10-136, RQ Radioactive material , low specific activity (LSA-II) fissile
excepted,7,UN3321,09/28/2010
Shipment No.10-148, Radioactive material, excepted package-limited quantity of material, 7,
UN2910, 10-21-10
Crystal River Unit 3 -2008 Radioactive Effluent Release Report, Table 9, Solid Waste and
Irradiated Fuel Shipments
Crystal River Unit 3 -2009 Radioactive Effluent Release Report, Crystal River Unit 3 -2009
Radioactive Effluent Release Report
RADMAN Database Report for Crystal River 3 Nuclear Plant, Change 146
2010 WDT-6 Primary Resin Sample Data Set Evaluation, 9/8/2010
CAP Documents
Nuclear Condition Report (NCR) No. 00419558, AB sump level rate of rise lowered with DW
secured
NCR No. 00415776, Black substance found on ground at rusty building
NCR No. 00416921, Boric Acid removal without oversight from nuclear waste technician or
radwaste supervision
Section 4OA1: Performance Indicator Verification
Procedures, Guidance Documents and Manuals
REG-NGGC-0009, NRC Performance Indicators and Monthly Operating Report Data,
Rev. 10
Records and Data Reviewed
2009 DRDE Evaluations Log
2009-2010 DRD Alarms Logs
Crystal River Unit 3 -2008 Radioactive Effluent Release Report, April 21, 2009
Attachment
7
Crystal River Unit 3 -2009 Radioactive Effluent Release Report, April 29, 2010
Release Permit 100138.006, 838.L SDT-1 Release to the Discharge Canal
Release Permit 100137.001.721.L WDT-10A Release to the Discharge Canal
Release Permit 100041.020.571.G Auxiliary Building Ventilation Continuous Release
Release Permit 100044. 018.063. G Reactor Containment Building Purge Batch Release
CAP Documents
AR 00359577, An individuals electronic dosimeter (ED) audio alarm faulted with a low battery
condition
4OA5: Other Activities
Steam Generator Replacement Inspection (IP50001)
EC 75220, Reactor Building Delamination Repair Phase 4 - Concrete Placement, Revision 21.
EC 75221, Reactor Building Delamination Repair Phase 5 -Retensioning, Revision 0.
Work Packages (WPs) 3-3732A, 3-3732B, 3-3732C, and 3-3732D, Restoration of Containment
Concrete Wall.
SGR EC Closure Schedules
Progress Energy SP-178, Containment Leakage Test-Type A Including Liner Plate
SGT Nonconformance Report (NCR) 1104, Rebar Spacing Did Not Meet the Requirements at
EL. 201 to 206
SGT NCR 1109, #11 Vertical Rebar Spacing at Tendon 34V21 Greater Than Allowable
SGT NCR 1115, Rebar Spacing Did not Meet the Requirements at EL. 216 to 221
Operator Licensing Training and Qualification Effectiveness Inspection
Operator Licensing Training and Qualification Effectiveness Inspection Records
Quick Hit Self-Assessment 421775, CR3 Operations Proficiency Training in Preparation for S/U
From R16, Attachment 8 to CAP-NGGC-0201, 09/28/2010.
Quick Hit Self-Assessment 422394, Restart Readiness, Attachment 8 to CAP-NGGC-0201,
09/17/2010.
Licensed Operator Requalification Training Schedules for Cycles 10A and 11A.
Licensed Operator Continuing Training (LOCT) Attendance Tracking Records (covering 12
training cycles).
Simulator Crew and Individual Evaluation Summaries for SES-69, Crew C, 10/21/2010.
Lesson Plans
OPS-4-54, Decay Heat Removal System, Revision 14, 02/11/2010.
OPS-5-1111, Post R16 Implementation Review for the Steam Generator Replacement (SGR)
and Extended Power Uprate (EPU) Projects, Revision 0, 10/08/2010.
OPS-5-1096, R16 Plant Walkdown, Revision 1, 10/15/2010.
OPS-5-1110, R16 Startup Sequence Overview and Testing Review Training, Revision 0,
10/15/2010.
OPS-5-1105, NGG Leadership Behaviors, Revision 0, 10/05/2010.
OPS-9-3083, Startup & Shutdown JIT Training, Revision 5, 12/21/2009.
OPS-9-3320, Decay Heat System Operations, Revision 0, 06/01/2010.
Attachment
8
Procedures
OP-202A, Refueling Outage Plant Heatup and Startup, Revision 18.
OP-204A, Plant Startup and Power Operations After R16/R17, Revision 3.
OP-304. Soluble Poison Concentration Control, Revision 29.
TRN-NGGC-0420, Nuclear Generation Group Standard Procedure: Conduct of Simulator
Training and Evaluation, Revision 0.
Simulator Dynamic Scenario Packages
OPS-9-3326 Scenario #2: MFWP Trip and Runback, HD Leak Causes Rapid Power Reduction
and Turbine Trip, Revision 1, 10/21/2010.
OPS-9-3326 Scenario #3: Turbine Bypass Valve Fails Open, RCP seal failure, RCP trip and
runback, Spurious Reactor Trip, Revision 1, 10/21/2010.
OPS-9-3326 Scenario #4: ICS Malfunction, Heater Drain System Malfunction, Steam Generator
Tube Rupture, Revision 1, 10/21/2010.
SES-69: Gland Steam Oscillations, High Main Generator Temperatures, ICS Neutron Error
Malfunction, Steam Generator Tube Leak with Complications, Revision 0, 10/04/2010.
NRC Temporary Instruction (TI) 2515/177, Managing Gas Accumulation in Emergency Core
Cooling, Decay Heat Removal, and Containment Spray Systems (NRC Generic Letter (GL) 2008-01)
Licensing Bases Documents
ML081330239, Crystal River Unit 3 - Three Month Response to NRC Generic Letter 2008-01,
Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and
Containment Spray Systems, May 8, 2008
ML082890555, Crystal River Unit 3 - Nine Month Response to NRC Generic Letter 2008-01,
Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and
Containment Spray Systems, October 13, 2008
ML102860131, Crystal River Unit 3 - Nine Month Supplemental (Post-Outage) Response to
NRC Generic Letter 2008-01, Managing Gas Accumulation in Emergency Core Cooling, Decay
Heat Removal, and Containment Spray Systems, October 8, 2010
ML100270031, Crystal River Unit 3 - Generic Letter 2008-01 Managing Gas Accumulation in
Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems, Response to
a Request for Additional Information, January 25, 2010
Miscellaneous
CAP-NGGC-0201-8-14, Quick Hit Self-Assessment Report - Assessment Number: QRPT
00397728, Rev. 14
EC 71569, High Point Vent Valves - Master (GL 2008-01 Outside the RB), Rev. 5
EC 72034, MUV-654, MUV-655, and MUV-657 High Point Vent Valves - Child (Master EC 71569), Rev. 1
Letter to R. Dulaney from LOCA Integrated Services (Westinghouse), Subject: PWROG Position
Paper on Non-condensable Gas Voids in ECCS Piping; Qualitative Engineering Judgment of
Potential Effects on Reactor Coolant System Transients Including Chapter 15 Events, Task 3 of
PA-SEE-450, Reference number: LTR-LIS-08-543, August 19, 2008
OP-103B, Plant Operating Curves, Rev. 40
Attachment
9
WO 01423288-01, GL2008-01; Perform Hanger Adjustment
WO 01423288-04, GL2008-01; Perform Hanger Adjustment
Drawings
FD-302-641, Decay Heat Removal, Rev. 81
FD-302-661, Make-up & Purification, Rev. 85
FD-302-711, Reactor Building Spray, Rev. 68
P-304-662, Make-up & Purification System Plan at EL. 95-0 Auxiliary Building, Rev. 20
PI-305-858, From MUV-69 to Make-up pump suctions 3A-3B-3C, Rev. 1
PI-305-811, GL 2008-01 Walkdown Schematic BS System
PI-305-810, GL 2008-01 Walkdown Schematic BS System
PI-305-815, GL 2008-01 Walkdown Schematic DH System A Train
Calculations
M09-0049, Evaluation of Gas Accumulation in CR3 LPI/DHR Suction Piping, Rev. 1
M09-0051, Evaluation of Gas Accumulation in CR3 DH and ECCS Discharge Piping, Rev. 1
M94-0053, Allowable MUT-1 Indicated Overpressure vs. Indicated Level, Rev. 7
NAI-1459-001, Comparison of GOTHIC Gas Transport Calculations with Test Data, Rev. 0
Action Requests Reviewed During Inspection
262601, SER 2-05 Revision 1 Gas Intrusion in Safety Systems
263132, Implementation Plan for GL 2008-01
284159, Track CR-3 Gas Accumulation Project Plan Actions Assignment Numbers 1 through 41
293769, Voiding Found in A BS Train of Discharge Side of BSP-1A
298140, Air Found in Abandoned Tee in MUP Discharge Header during UT
302656, OE27625 - RHR System Inoperability in Modes 3 and 4
312009, UT Examination of A DH Discharge Pipe Reveals Voiding
315347, UT Examination of A DH Discharge Pipe Reveals Minor Void
344364, Common Suction to Make-up Pumps had Minor Gas Pockets
346131, Existing Acceptable Gas Voids MUP Common Suction from MUV-60
358532, Gas/Void Found at MUV-37
358790, MUV-42 Upstream Elbow Just Over 50% Full Based on UT
365715, NSAL-09-8 Presence of Vapor in Emergency Core Cooling System
382171, Void Pocket Observed in DH Dropline during UT Inspection
417759, Fill and Vent Sequence for B DH Train
Procedures
EGR-NGGC-0005, Engineering Change, Rev. 31
EOP-6, Steam Generator Tube Rupture, Rev. 20
NDEP-0438, Ultrasonic Procedure for Determination of Liquid Level in Components
SP-630, MU/HPI Check Valves Full Flow Test, Rev 21
SP-340B, DHP-1A, BSP-1A and Valve Surveillance, Rev 63
SP-340E, DHP-1B, BSP-1B and Valve Surveillance, Rev 43
OP-402, Makeup and Purification System, Rev 159
OP-404, Decay Heat Removal System, Rev 155
SP-435, Valve Testing During Cold Shutdown, Rev 68
Attachment
10
Completed Testing
WO 01426356-01, DHV-41/5/6 GL2008-01; UT Inspection
WO 01627233-01, DHV-41, GL2008-01; UT Inspection
WO 01627234-01, DHV-6, GL2008-01; UT Inspection
WO 01663114-01, DHV-5, GL2008-01; UT Inspection
WO 01679519-01, DHV-41, GL2008-01; UT Inspection
WO 01679520-01, DHV-6, GL2008-01; UT Inspection
WO 01713641-01, DHV-5, GL2008-01; UT Inspection
WO 01725959-01, DHV-41, GL2008-01; UT Inspection
WO 01725960-01, DHV-6, GL2008-01; UT Inspection
WO 01757745-01, DHV-5, GL2008-01; UT Inspection
WO 01767042-01, DHV-6, GL2008-01; UT Inspection
NCRs Generated As a Result of Inspection
429660, During NRC Inspection for GL 08-01 Ziplevel Uncertainty
429889, NRC GL 08-01 Inspection Identified WO Discrepancies
429950, GL 08-01 NRC Inspection Identified DH ISO C/D Not Evaluated
430217, Suspect Information Found on VT-3 Document for MUH-518
430255, NRC GL 08-01 Inspection Identifies Need for LER Review
430264, NRC GL 08-01 Inspection Identified QH-SA Deficiency
Attachment